ML20216G612
| ML20216G612 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 04/13/1998 |
| From: | Stetka T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | Dugger C ENTERGY OPERATIONS, INC. |
| References | |
| 50-382-97-25, EA-98-022, EA-98-22, NUDOCS 9804200401 | |
| Download: ML20216G612 (150) | |
See also: IR 05000382/1997025
Text
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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REGION IV
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611 RYAN PLAZA DRIVE, SulTE 400
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April 13,1998
EA 98-022
Charles M. Dugger, Vice President
Operations - Waterford 3
Entergy Operations, Inc.
P.O. Box B
Killona, Louisiana 70066
SUBJECT: March 26,1998, PREDECISIONAL ENFORCEMENT CONFERENCE SUMMARY
FOR NRC INSPECTION REPORT 50-382/97-25
Dear Mr. Dugger;
i
This refers to the meeting with you and your staff conducted in the Region IV office on
'
March 26,1998. The meeting was conducted with those listed on the attendance list
(Enclosure 1)in accordance with the enclosed agenda (Enclosure 2). This predecisional
enforcement conference was convened to discuss six apparent violations. These involved two
apparent violations of 10 CFR 50.46, " Acceptance Criteria for Emergency Core Cooling Systems
for Light Water Nuclear Power Reac+ ors"; two apparent violations of 10 CFR 50.59, " Changes,
Tests, and Experiments"; and one apparent violation of 10 CFR Part 50, Appendix B,
Criterion XVI, " Corrective Action," and Criterion XI, " Test Control." During the meeting, the NRC
and your staff discussed the apparent violations, which are enclosed (Enclosure 3). The NRC
also requested that your staff discuss related issues and provide comments on the inspection
report. Your staff presented their views regarding these issues. Your handout, which provided a
basis for these discussions, is also included (Enclosures 4).
At the close of the meeting, you provided a written summary of your position on " Determining A
Margin of Safety for Identifying an Unreviewed Safety Question Per 50.59." You also provided
i
written comments on the inspection report, which you stated were minor and would not
materially impact the enforcement. Both of these documer.t. are enclosed (Enclosuies 5 and 6).
You will be advised by separate correspondence of the results of our deliberations regarding the
apparent violations and the issues discussed during the conference. No response regarding
'
these apparent violations is required at this time.
In accordance with Section 2.790 of the NRC's " Rules of Practice," Part 2 Title 10, Code of
'
Federal Regulations, a copy of this letter will be placed in the NRC's Public Document Room.
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9804200401 980413
ADOCK 05000382
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Entergy Operations, Inc.
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Should you have any questions conceming this matter, we will be pleased to discuss them with
you.
Sincerely,
Thomas F. Stetka, Acting Chief
Engineering Branch
Division of Reactor Safety
Docket No.:
50-382
License No.: NPF-38
Enclosures:
1. Attendance List
2. Agenda
3. Apparent Violations
4. Licensee Presentation
5. Licensee Paper: " Determining a Margin of Safety for Identifying an Unreviewed Safety
Question Per 50.59"
6. Licensee inspection Report Comments
i
cc w/ Enclosures 1-3:
Executive Vice President and
Chief Operating Officer
Entergy Operations, Inc.
P.O. Box 31995
Jackson, Mississippi 39286-1995
Vice President, Operations Support
Entergy Operations, Inc.
P.O. Box 31995
Jackson, Mississippi 39286-1995
Wise, Carter, Child & Caraway
P.O. Box 651
Jackson, Mississippi 39205
General Manager, Plant Operations
Waterford 3 SES
Entergy Operations, Inc.
P.O. Box B
Killona, Louisiana 70066
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Entergy Operations, Inc.
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Manager- Licensing Manager
Waterford 3 SES
Entergy Operations, Inc.
P.O. Box B
Killona, Louisiana 70066
Chairman
Louisiana Public Service Commission
One American Place, Suite 1630
Baton Rouge, Louisiana 70825-1697
Director, Nuclear Safety &
Regulatory Affairs
Waterford 3 SES
Entergy Operations, Inc.
P.O. Box B
Killona, Louisiana 70066
s.
William H. Spell, Administrator
Louisiana Radiation Protection Division
P.O. Box 82135
Baton Rouge, Louisiana 70884-2135
Parish President
St. Charles Parish
P.O. Box 302
Hahnville, Louisiana 70057
Mr. William A. Cross
Bethesda Licensing Office
3 Metro Center
Suite 610
Bethesda, Maryland 20814
Winston & Gtrawn
1400 L Street, N.W.
Washington, D.C. 20005-3502
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MIS System
Project Engineer (DRP/D)
RIV File
Branch Chief (DRP/TSS)
Lieberman, OE (0-7H5)
OE:EA File (0-7H5)
G. Sanborn
DOCUMENT NAME: R:\\_WATTWT725MS.LJS
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PREDECISIONAL ENFORCEMENT CONFERENCE ATTENDANCE
LICENSEE / FACILITY
Entergy Operations, Inc.
Waterford-3
DATE/ TIME
March 26,1998, @ 1 p.m. (CST)
CONFERENCE LOCATION
Region IV, Training Conference Room
Arhngton, TX
EA NUMBER
EA 98-022
LICENSEE REPRESENTATIVES
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Waterford-3
DATE/ TIME
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CONFERENCE LOCATION
Region IV, Training Conference Room
Arlington, TX
EA NUMBER
EA 98-022
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PREDECISIONAL ENFORCEMENT CONFERENCE AGENDA
CONFERENCE WITH ENTERGY OPERATIONS, INC.
MARCH 26,1998
)
NRC REGION IV, ARLINGTON, TEXAS
4
1.
INTRODUCTIONS / OPENING REMARKS - REGIONAL
ADMINISTRATOR
2.
ENFORCEMENT PROCESS - ENFORCEMENT OFFICER
3.
APPARENT VIOLATIONS & REGULATORY CONCERNS -
DIRECTOR, DRS
4.
LICENSEE PRESENTATION -
5.
BREAK (10-MINUTE NRC CAUCUS IF NECESSARY)
6.
RESUMPTION OF CONFERENCE
7.
CLOSING REMARKS - LICENSEE
8.
CLOSING REMARKS - REGIONAL ADMINISTRATOR
.
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,
APPARENT VIOLATIONS *
PREDECISIONAL ENFORCEMENT CONFERENCE
]
i
ENTERGY OPERATIONS, INC.
MARCH 26,1998
- NOTE: THE APPARENT VIOLATIONS DISCUSSED AT THIS PREDECISIONAL
ENFORCEMENT CONFERENCE ARE SUBJECT TO FURTHER REVIEW AND MAY BE
REVISED PRIOR TO ANYRESULTING ENFORCEMENT ACTION.
.
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APPARENT VIOLATION
1.
10 CFR Part 50.46 (a)(3)(1) requires, in part, that each holder of an operating license
shall estimate the effect of any change to or error in an acceptable emergency core
cooling system (ECCS) evaluation model or in the application of such a model to
determine if the change results in a calculated peak fuel cladding temperature that is
different bv more than 50*F from the temperature calculated for the limiting transient
using the latest acceptable ECCS model.
Contrary to the above, on December 5,1997, the ECCS evaluation model was found to
be in error, in that available flow for the hich pressure safety injection (HPSI) system, an
ECCS system, was less than assumed. Specifically, the evaluation model for a small
break loss-of-coolant accident assumed that 621.8 gpm of HPSI flow would be available
to cool the core. However, after test instrument uncertainty was considered, only 599
gpm of HPSI flow was found to be available. The estimate of the effect of this HPSI flow
reduction on calculated peak fuel cladding temperature was not made until Decembar
18,1997,
i
a
MAYBE REVISED
1
i
APPARENT VIOLATION
2,
10 CFR Part 50.46 (a)(3)(ii) requires, "For each change to or error discovered in an
acceptable ECCS evaluation model or in the application of such a model that affects the
temperature calculation, the applicant shall report the nature of the change or error and
its estimated effect on the limiting emergency core cooling system (ECCS) analysis to
the Commission at least annually as specified in 10 CFR 50.4. If the change or error is
significant, the applicant shall provide this report within 30 days and include with the
report a proposed schedule for providing a reanalysis or taking other action as may be
needed to show compliance with 10 CFR 50.46.*
10 CFR Part 50.46 (a)(3)(ii) further requires, "Any change or error correction that results
in a calculated ECCS performance that does not conform to the criteria set forth in
paragraph (b) of this section is a reportable event as described in . . 10 CFR Part 50.72
and 10 CFR Part 50.73."
10 CFR Part 50.46 (b)(1) requires, "The calculated maximum fuel element cladding
temperature shall not exceed 2200*F."
10 CFR 50 Part 72 (b)(ii)(B) requires, in part *the licensee shall notify the NRC as soon
as practical and in all cases within one hour of the occurrence of any of the following:. . .
(ii) Any event or condition during operation that results in . . the nuclear power plant
being:. . . (B) In a condition that is outside the design basis of the plant."
Contrary to the above:
a.
As of January 22,1998, a proposed schedule for a limiting ECCS reanalysis,
which corrected the significant high pressure safety injection (HPSI) system flow
error or for taking other action as may be needed to show compliance with 10 CFR Part 50.46, was not provided.
b.
On December 5,1997, an error correction, which would have resulted in a
calculated ECCS performance that did not conform to the criteria set forth in
paragraph (b) of 10 CFR 50.46 was not reported within one hour. Specifically,
using the licensing basis analysis and the available HPSI flow, it was identified
that available HPSI flow was less than that stated in the safety analysis and that
peak fuel cladding temperature would have exceeded 2200 F, a condition outside
the design basis of the plant. The condition was not reported until December 18,
1997.
.
THIS APPARENT VIOLATION IS SUBJECT TO FURTHER REVIEW AND
MAY BE REVISED
APPARENT VIOLATION
3.
10 CFR Part 50, Appendix B, Criterion XVI, " Corrective Action," requires that measures
shall be established to ensure that conditions adverse to quality, such es failures,
malfunctions, deficiencies, deviations, defective material and equipment, and
noncenformances are promptly identified and corrected.
Contrary to the above, a condition adverse to quality was identified but was not promptly
corrected. On July 16,1996, a self assessment noted that neither Surveillance
Procedure OP-903-108, "Si Flow Balance Test," Revision 3, Change 1, nor Technical Specification 4.5.2.h included an adequate allowance for flow measurement test
instrument uncertainty. As of December 18,1997, neither document had been revised or
amended to address this deficiency.
.
THIS APPARENT VIOLATION IS SUBJECT TO FURTHER REVIEW AND
MAY BE REVISED
i
!
APPARENT VIOLATION
4.
10 CFR Part 50, Appendix B, Criterion XI, requires, in part, that all testing required to
demonstrate that structures, systems, and components will perform satisfactorily in
service is performed in accordance with written test procedures, which incorporate the
requirements and acceptance limits contained in applicable design documents.10 CFR Part 50, Appendix B, Criterion XI, further requires, that test procedures shall include
provisions for assuring that adequate test instrumentation is used.
Surve;llance Procedure OP-903-108, "SI Flow Balance Test," Revision 3, Change 1,
provides instructions for performing the flow balance of the high pressure safety injection
(HPSI) system that is required by Technical Specification 4.5.2.h. The bases section for
Technical Specification 3/4.5.2 states that the surveillance requirements ensure that, at a
minimum, the assumptions used in the safety analysis are met. In addition, Technical Specification 4.5.2.g required the verification of the correct position of each electrical
and/or mechanical position stop for the emergency core cooling system (ECCS) throttle
valves each time the valve was cycled. Surveillance Procedure OP-903-010, "ECCS
Throttle Valves Position Verification," Revision 3, implemented this ted- ical specification
requirement and allowed a +/- 2 percent tolerance band for the as-found flow control
valve position from its set point value.
Engineering Procedure UNT-001-002,"ACCW & CCW System Flow Balance," Revision
14, provides instructions for performing a flow balance of the auxiliary component cooling
water (ACCW) and the component cooling water (CCW) systems. This test ensures the
systems are capable of achieving the flows assumed in the thermal performance
analysis.
Contrary to the above:
.
a.
From April 10,1994, until December 18,1997, Surveillance Procedure
OP-903-108 did not include provisions for assuring that adequate test
instrumentation was used. Specifically, the minimum flow of 675 gpm required by
Technical Specification 4.5.2.h included an allowance of 5 gpm/ leg, to account for
flow instrument measurernent uncertainty. However, Surveillance Procedure
OP-903-108 directed personnel to use flow instruments which had a flow
measurement uncertainty of approximately 18 gpm/ leg.
,
THIS APPARENT VIOLATION IS SUBJECT TO FURTHER REVIEW AND
MAY BE REVISED
APPARENT VIOLATION
b.
From April 10,1994 until December 18,1997, Surveillance Procedure
OP-903-108 did not adequately incorporate the requirements and acceptance
limits contained in Technical Specification 4.5.2.h, Surveillance Procedure OP-
903-010, and the safety analysis. Specifically, the acceptance limit for flow in
Procedure OP-903-108 did not include an allowance for throttle valve position
variability allowed by Procedure OP-903-010. Consideration of this allowance
was necessary to ensure that, for the worst case ECCS throttle valve position, the
flow assumptions used in the safety analysis would be met.
c.
As of December 18,1997, Engineering Procedure UNT-001-002 did not include
acceptance limits, which addressed all applicable information in design
'
documents. Specifically, the test acceptance limit did not include an allowance
for flow instrument uncertainty, which was necessary to ensure that, at a
minimum, the flow assumptions used in the ACCW/CCW heat exchanger thermal
performance analysis was met.
i
.
THIS APPARENT VIOLATION IS SUBJECT TO FURTHER REVIEW AND
MAY BE REVISED
APPARENT VIOLATION
5.
10 CFR 50.59(b)(1) states, in part, that the licensee shall maintain records of changes in
the facility and of changes in procedures made pursuant to this section to the extent that
these changes constitute changes in the facility as described in the safety analysis report
or changes in procedures as described in the safety analysis report. These records must
include a written safety evaluation which provides the bases for the determination that
the change does not involve an unreviewed safety question.
Updated Final Safety Analysis Report (UFSAR) Section 6.3.3.3.1, " Safety injection
Assumptions," states that in addition to the safety injection system flows, injection flow
from the charging pumps is credited in the small break loss-of-coolant accident analysis.
Procedure OP-902, " Loss of Coolant Accident Recovery Procedure," Revision 7,
Change 2 was the general loss-of-coolant accident recovery procedure, which included
responses to small- and large-break loss-of-coolant accidents.
Contrary to the above, from July 6,1997 to December 18,1997, a record of a written
safety evaluation was not maintained for Procedure OP-902, Revision 7, Change 2, in
that this procedure was changed to instruct operators to secure the charging pumps,
during a loss-of-coolant accident (LOCA) recovery, when a recirculation actuation signal
occurs. This change constituted a change to the procedure as described in UFSAR
Section 6.3.3.3.1.
.
THIS APPARENT VIOLATION IS SUBJECT TO FURTHER REVIEW AND
MAY BE REVISED
o
1
1
APPARENT VIOLATION
6.
10 CFR 50.59(a)(1) states, in part, that a licensee may make changes in the facility as
described in the safety analysis report and changes in procedures as described in the
safety analysis report without prior Commission approval unless the proposed change
involves a change in the technical specifications incorporated in the license or an
unreviewed safety question.
10 CFR 50.59(a)(2) states, that a proposed change, test, experiment shall be deemed to
involve an unreviewed safety question (1) if the probability of occurrence or the
consequences of an accident or malfunction of equipment important to safety previously
evaluated in the safety analysis report may be increased; or (ii) if a possibility for an
accident or malfunction of a different type than any evaluated previously in the safety
analysis report may be created or (iii) if the margin of safety as defined in the basis for
any technical specification is reduced.
a.
Technical Specification Basis 3/4.5.2 states that the surveillance requirements
ensure that, at a minimum, the assumptions used in the safety analysis are met.
The basis further states that maintenance of proper flow resistance and pressure
drop in the piping system to each injection point is necessary to provide an
acceptable level of total emergency core cooling system (ECCS) flow to all
injection points equal to or above that assumed in the ECCS loss-of-coolant
(LOCA) analyses.
UFSAR Table 6.3-7, "HPSI [high pressure safety injection) Pump Minimum
Delivered Flow To RCS [ reactor coolant system] For The 0.04 FT2 Break
Analysis," specifies a minimum delivered flow of 207.25 gpm.
Contrary to the above, the licensee approved a change to the facility as described
in UFSAR Table 6.3-7, which involved an unreviewed safety question, without
)
obtaining prior Commission approval. Specifically, on September 18,1996, the
licensee approved Revision 1 of Calculation EC-195-011,"SI-HPSI Flow
instrumentation Loop Uncertainty Calculation." After consideration of the results
of Calculation EC-195-011, a HPSI system that had been flow balanced to satisfy
i
Technical Specification 4.5.2.h could only be relied on to supply 197 gpm in one
injection path, which was less than the 207.25 gpm specified in UFSAR Table
6.3-7. Since Technical Specification 4.5.2.h no longer ensured an acceptable
level of flow to all points equal to or above that assumed in the currently licensed
,
ECCS-LOCA analyses, the change reduced the margin of safety defined in the
basis for Technical Specification 3/4.5.2.
.
THIS APPARENT VIOLATION IS SUBJECT TO FURTHER REVIEW AND
MAYBE REVISED
APPARENT VIOLATION
b.
From December 18,1984, until July 10,1997, Technical Specification Bases 3/4.7.1.2
stated: *Each electric-driven emergency feedwater pump is capable of delivering a total
feedwater flow of 350 gpm at a pressure of 1163 psig to the entrance of the steam
- generators. The steam-driven emergency feedwater pump is capable of delivering a
total feedwater flow of 700 gpm at a pressure of 1163 psig to the entrance of the steam
generators."
Until July 10,1997, UFSAR Section 10.4.9.2, " Emergency Feedwater System
Description," stated that the turbine driven pump or both motor-driven pumps together
have been designed to provide 700 gpm flow to the steam generators upon loss of
feedwater flow in order to remove decay heat and to reduce reactor coolant system
temperature and pressure to the shutdown cooling entry conditions.
UFSAR Section 15.2.3.1.2, "[Feedwater System Pipe Breaks] Sequence of Events and
Systems Operation," states that operator action is not required until 30 minutes after the
event.
UFSAR Section 15.6.3.4 includes an estimate of the frequency of an inadvertent opening
of a pressurizer safety valve.
Contrary to the above, on July 10,1997, the licensee approved a change to the facility
as described in the UFSAR, which involved an unreviewed safety question, without prior
Commission approval. Specifically, Safety Evaluation 97-165 for Licensing Document
Change Request (LDCR) 97-0034, revised Technical Specification Bases 3/4.7.1.2 to
reduce the emergency feedwater pump capability requirements. The revised basis
stated that: "The two electric-driven emergency feedwater pumps combined are capable
of delivering a total feedwater flow of 575 gpm at a pressure of 1102 psig to the entrance
of the steam generators. The steam-driven emergency feedwater pump is cE:pable of
delivering a total feedwater flow of 575 gpm at a pressure of 1102 psig to the entrance of
the s'.eam generator."
1.
The reduction in the ernergency feedwater pump capability requirements below
those specified in UFSAR Section 10.4.9.2, and below the values assumed in the
safety analysis, resulted in a reduction in the margin of safety as defined in the
basis for Technical Specification 3/4.7.1.2.
2.
The reduction in the emergency feedwater pump capability requirements also
resulted in a reduction in the time available for operator action to prevent filling
the prassurizer after a main feedline break. The time available for operator action
was reduced from 30 minutes to between 7 to 25 minutes, depending on
analysis techniques. The reduction in operator action time increased the
probability of a previously evaluated malfunction (stuck-open pressurizer relief
valves) and increased the probability of an accident (a loss-of-coolant accident).
THIS APPARENT VIOLATION IS SUBJECT TO FURTHER REVIEW AND
MA'(BE REVISED
APPARENT VIOLATION
c.
UFSAR Table 6.2-32, " Containment Penetrations and Isolation Valves," states
that Hydrogen Analyzer Valves HRAISV0110A(B), HRAISV0109A(B), and
HRAISV0126A(B) are normally locked closed. The table further states that the
valves are open post-accident and they receive a containment isolation actuation
signel with a maximum closure time of 5 seconds.
]
)
Contrary to the above, on November 28,1997, a change was made to the facility
)
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as described in UFSAR Table 6.2-32 which involved an unreviewed safety
question, without prior Commission approval. Specifically, LDCR 98-0032,
reclassified Hydrogen Analyzer Valves HRAISV0110A(B), HRAISV0109A(B), and
HRAISV0126A(B) from automatic to manual! remote manual operation. This
reclassification was in conflict with UFSAR Table 6.2-32. The reclassification
introduced the possibility of a malfunction of a different type than any previously
evaluated in the safety analysis report (i.e., failure to isolate the containment
isolation valves within 5 seconds).
~
THIS APPARENT VIOLATION IS SUBJECT TO FURTHER REVIEW AND
MAYBE REVISED
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4
DETERMINING A MARGIN OF SAFETY FOR IDENTIFYING
AN UNREVIEWED SAFETY QUESTION PER 50.59
)
Background:
The original 10CFR 50.59 rule developed in 1962 did not include the Margin of Safety
question as currently written in the rule. The changes to Q 50.59 were issued in December 1968
i
in the same Federal Register notice that issued substantial changes to 10 CFR 50.34 and 10 CFR 50.36. These changes had been proposed over two years earlier in recognition of the fact
that technical specifications could serve their purpose with much less detail. To accomplish
this, bases for the technical specifications were required and the contents of safety analysis
reports were more explicitly specified including new emphasis on components. These changes
in 10 CFR 50.34 and 10 CFR 50.36 resulted in the addition of the phrases " malfunction of
equipment important to safety" and " margin of safety as defined in the basis for any technical
specification" to the definition of unreviewed safety question. The margins of safety which
,
were contained in the Bases of the technical specifications were considered important and were
not to be reduced without explicit NRC review and approval. All of these changes were
proposed in a single Federal Register notice (Vol. 31, No.158, pp.10891-10894) and
implemented in a single Federal Register notice (Vol. 33, No. 244, pp.18610-18612).
In
development of NSAC-125, the " basis" was conservatively expanded to include not only the
Bases section of the Technical Specifications, but other locations that contained information
that would establish a margin of safety such as the Safety Analysis Report.
With the improvement in the technical specifications in late 1968, the Margin of Safety
question was developed to ensure that specific safety margins were not changed without NRC
involvement and review. It was the NRC intent that most Bases information would no longer
be elevated to the same level of control and importance as that contained in the Technical
Specifications. However, certain controlling safety margins that could ultimately impact plant
safety needed to be maintained to avoid degradation of the plant safety margins and the
Technical Specifications. It needs to be recognized that the NRC review and approval of an
unreviewed safety question under 10CFR50.90, elevates the regulatory action to the same level
as that contained in the Operating License or the Technical Specifications. Therefore, the NRC
clearly intended that the identification of a Margin of Safety through a USQ would only be a
relatively small number of unique safety margins that would impact the ultimate safety of the
plant. Otherwise the technical specifications would include additional values and parameters
which are relegated to the Bases of the Technical Specifications. Clearly, the NRC intended
that
the values written in the Tech Spec Bases should not be considered to represent a Margin of
Safety.
The Margin of Safety determination in NSAC-125, Guidelines for 10 CFR 50.59 Safety
Evaluations, was developed in 1989 with the above perspective in mind. Even though never
accepted by the NRC staff, the development of NSAC-125 (now NEI 96-07, Rev.1) was also
based on NRC input to an early draft of NSAC-125. The NRC's comments and approach to
1
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3/20/98
Margin of Safety was best established in correspondence from Charles Rossi, Dir., Div. of
Operational Events Assessment to Thomas Tipton, NUMARC, dated May 10,1989. Most of
the comments provided by the NRC were incorporated into the final revision of NSAC-125.
Waterford's 50.59 Program including the approach to Margin of Safety is based on the
philosophy contained in NSAC-125.
Determination of Various Plant Margins
There are various plant margins associated with any change. These include design margins,
j
operating margins and safety margins. This is recognized by the NRC staffin the 1989 NRC
letter to NUMARC which states:
"ne staff recognizes that there are " margins" associated with SAR analyses to account for uncertainties
in the design, construction, and operation of a nuclear power plant (e.g., conservatisms in computer
'
modeling and codes, allowances for instrument drift and system response time). Dese " margins" may be
reduced by licer. sees without prior staff approval provided the specific acceptance conditions, criteria and
,
limits (including models, tests, uncertainties, penalties, methodology, etc.) are not adversely affected or
,
I
invalidated." [ Incorporated into NSAC-125].
Entergy Operations has informally defined the safety, design and operating margins in order of
hierarchy as follows:
Safety Margin - One of the criteria that defines an unreviewed safety question in 10 CFR 50.59 is "if the
margin of safety as defined in the bases for any technical specifications is reduced." The Technical
Specification " margin of safety" is the region (difference) between the NRC acceptance limit and the
regulatory limit (or failure point) of the (fission product) barrier. De acceptance limit is the value
'
approved by the NRC as part of the licensing basis. A change which results in a parameter exceeding its
acceptance limit would be an unreviewed safety question. The failure point is simply the value of
parameters at which a protective boundary (fission product barrier) would probably fail. A change made
directly to a protective boundary requires evaluation for the potential to reduce the margin of safety.
[ Example: A change is made that directly or indirectly modifies RCS flows, temperatures or pressures
where RCS peak pressure is the safety limit or protective boundary potentially afTected. The resulting
analysis changes the peak RCS pressure from 2400 psig to 2450 psig. If the NRC had specifically accepted
2400 psig as the peak pressure, even though allowed to go to 2750 psig per ASME (110% design pressure),
this would be a change in the Margin of Safety and would represent a USQ. Ilowever, if the NRC stated
that this is below the maximum design pressure of 2500 psig, then no Margin of Safety is impacted. In the
.
second case, the additional 50 psig increase would be design margin.]
Design Margin - Design margins include design inputs which are based on the functions and performance
requirements of the structure, system and component, the applicable codes and standards, and the range of
i
design and operating conditions applicable to the design. Design inputs generally include allowances for
various uncertainties, such as calculation inaccuracies, manufacturing tolerances, instrument uncertainties
and variatiot's in material properties. De designer may also include allowances to accommodate inservice
degradation, such as wear, corrosion, fouling and malfunctions. After a plant is in operation, as built and
operating performance data may justify reducing design allowances to values sufficient for repair and
replacement activities, creating margin that may be allocated for other purposes. [ Example: A change is
made that directly or indirectly modifies RCS flows, temperatures or pressures where RCS peak pressure is
the safety limit or protective boundary potentially affected. The resulting analysis does not change the peak
RCS pressure. Even though the RCS peak pressure had the potential to be affected, no change was
determined and, therefore, does not represent a reduction in a Margin of Safety. The change is considered
to affect design margin only.]
2
1
E
l
3/20/98
Operating Margin - The design output provides a perfonnarice range for operation (normal operating range)
to be maintained by the control systems and the human operator. The normal operating range must be far
.
enough from alarm setpoints, protection setpoints and operability limits to allow reliable operation. The
margin that exists between the nominal performance of the system or component and the alarm setpoints,
protection setpoints, and limiting conditions of operation can be considered as operating margin. The
i
operating margin must be large enough to accommodate normal variations in equipment and operator
performance. Ample operating margins enhance safety by reducing challenges to safety systems.
[ Example: : A procedure change is made that increases a plant setpoint in the field, but does r.ot modify
existing RCS flows, temperatures or pressures or established design setpoints. The change is within the
pre-established analyses of the system design and no change to design documentation is necessary. This
change would be considered a change in operating margin.]
The design and operating margins are within the control of the licensee and the safety margin is
within the control of the NRC for any reductions determined necessary.
Identification of a " Margin of Safety"
Based on the above, it is important to recognize during a change that if margin is being
reduced, whether it involves a design, operating or safety margin. The Margin of Safety by its
definition is only intended to be those upper level controlling safety limits or protective
boundaries that must be maintained to ensure that overall plant safety is not reduced. In
determining whether an unreviewed safety question exists for a Margin of Safety, Waterford
has established four (4) primary questions that need to be answered in responding to the Margin
of Safety question. These questions are based on NSAC-125 and are as follows:
1. If the change is to a protective boundary, discuss how the boundary is affected.
j
2. Identify the margins of safety (related to boundary performance) that may be
j
affected by the proposed change or activity.
3. Discuss how the accident response, as affected by the proposed change or activity,
relates to the appropriate acceptance limits.
4. If applicable, provide the results of an analysis that accounts for the proposed
change or activity and shows the impact on margin of safety.
A general flowchart of the Margin of Safety determination process for Waterford is provided as
Figure 1.
Il
l
if the change is to a protective boundary, discuss how the boundary is affected.
Per NSAC-125, a Margin of Safety would typically only be concemed with the impact of those
physical parameters that define the performance of safety limits or protective (fission product)
barriers. These parameters are as follows:
i
3
1
3/20/98
1
Fuel Cladding: DNBR/MCPR
Fuel Temperature
Fuel Enthalpy
_
Clad Strain
Clad Temperature
,
Clad Oxidation
RCS Boundary: Pressure, stresses (ASME Code for normal, upset,
emergency and faulted conditions)
Containment: Pressure
!
The May 10,1989 NRC letter to NUMARC also states:
"The Technical Specification Bases for safety limits on many plants are examples of Bases that
explicitly address margins of safety. Changes in transient or accident core thermal hydraulic condition
or peak reactor coolant pressure which do not violate the fuel design limits or reactor coolant system
design pressure in the Bases for the safety limits specified in the plant Technical Specification do not
constitute a reduction in the margin of safety as used in 10CFR50.59 provided the changes are made
consistent with previously accepted methods and specific acceptance conditions, criteria, and limits
(including models, test uncertainties, penalties, methods etc.)." [ Incorporated into NSAC-125 from NRC
May 10,1989 letter)
Therefore, the primary focus on determining a Margin of Safety involves only those upper level
plant safety limits or protective boundaries being impacted.
Any analysis margins not
impacting the safety limits or protective boundaries would generally involve a design or
operating margin and not a safety margin, RCS or ECCS flows, temperatures or pressures in
themselves do not represent Margins of Safety even though the analytical result of their change
can impact a Margin of Safety.
l
Identify the Margins ofSafety (related to boundary performance) that may be affected by
{
theproposed change or activity.
If a safety limit or protective boundary is potentially being impacted, then the Margin (s) of
Safety (explicit or implicit) need to be identified. Again per NSAC-125 this step is
determined as follows:
In making the judgment on whether the margin is reduced, the decision should be based on physical
parameters or conditions which can be observed or calculated. In some cases, the value of the parameter
identified as the acceptance limit is conservatively chosen with respect to the outer bound of NRC
acceptance. For example, the ASME Code limit for the RCS pressure boundary,110% of design
(typically 2750 psia) is often chosen as the acceptance limit, although the same confidence level in RCS
- pressure boundary integrity could be demonstrated up to a higher pressure value.
The margin may be implicit rather than being explicitly expressed as a numeri:al value. The precise
determination of a numerical value associated with a change is not always required to comply with 10 CFR 50.59. Implicit margins are conditions for NRC acceptance for a computer code, method, industry
accepted practice, or penalty. It may be sufficient to determine only the direction of the margin change
(i.e., increasing or decreasing). If the margin is reduced, the change, test, or experiment may involve an
4
3/20/98
USQ. A margin of safety defined in the bases section of the technical specifications may depend on a
parameter other than one of the process variables. [ Incorporated into NSAC-125 from NRC May 10
"
1989 letter)
Therefore, based on the above, the action would be to identify those explicit or implicit
margins of safety either by numerical value or by consideration of a system design feature
that would potentially impact a safety limit or protective boundary. This would be
represented in a change in the specific TS Bases /SAR analyzed values or parameters for
safety limits or protective boundaries (# change for DNBR, peak pressure change for
containment, peak pressure change for RCS, etc.). If there is no change or impact to a
previously analysed safety limit or protective boundary, then there is no change in the
Margin of Safety.
Discuss how the accident response, as affected by the proposed change or activity, relates
to the appropriate acceptance limits.
This step now requires the reviewer to make a comparison of the proposed change to determine
whether the Margin of Safety is affected which would prompt a USQ. As also discussed in
NSAC 125,
"The detennination of whether or not a reduction in margin is involved is based on the results of the
analysis and not on the change itself. For example, an increase in initial conditions (not already limited
by technical specifications) in the non-conservative direction can be compensated for by lowering a
setpoint or reallocating analysis conservatisms. If the analysis results continue to be bounded by the
acceptance limit, a reduction of margin is not involved. In this respect, the evaluation of reduction in
margin of safety is performed in a way analogous to the way changes to the LOCA analysis are evaluated
to determine if NRC review is required. The criterion for seeking prior review and approval is based on
the extent of the change in LOCA analysis results and not on the input change per se."
Next, the value of the acceptance limit is identified. This is the value at which the confidence level in the
integrity of the barrier decreases. The margin between the failure point and the acceptance limit is
regulated by the NRC.
The acceptance limit is that value approved by the NRC (e.g., ASME Code stress limits, a DNBR limit or
MCPR limit or containment design pressure). This limit, if exceeded in an application, would constitute
a need for prior NRC review. The acceptance limit that is reviewed and approved by the NRC as part of
the licensing " basis" is typically that value that is proposed by the licensee in the SAR and has been
modified by the NRC's SER. To find the acceptance limit, one must determine the original licensing
basis of the parameter in question. [ Incorporated into NSAC-125 from NRC May 10,1989 letter]
The licensee should apply the same methodology, with and without the proposed change, when
evaluating a change to determine its effect upon the margin of safety. For example, assume the licensee
performed, but did not submit to the NRC, analysis to determine the peak pressure for a system (perhaps
as a function of time) following an accident. Further, assume the licensee then asked the NRC to license
the plant based on a more conservative (bounding) limit or curve (because of analyses uncertainties), and
the NRC reviewed and approved the bounding limit or curve. In this case, the licensee could make a
change which would increase the peak system pressure provided a more precise analysis, with reduced
uncertainties,left the bounding limit or curve valid. However, if the specific methodology for computing
the bounding limit or curve or combining uncertainties (such as instrument errors) was submitted to the
5
-
3/20/98
. NRC in support of the licensing action, reductions in margin associated with this methodology would
['
constitute an IJSQ [ Incorporated into NSAC-125 from NRC May 10,1989 letter]
Therefore, if applicable, the specific acceptance limits from the NRC review are identified
and are evaluated against the resulting analysis p rformed by an accident analysis or other
design analysis. This establishes the basis on whether a USQ exists or not.
If applicable, provide the results of an analysis that accountsfor the proposed change or
activity andshows the impact on margin ofsafety.
This step merely provides documentation, if applicable, on the conclusion reached in the above
step for whether a USQ exists or not.
-
Conclusion
The Margin of Safety approach applied in the Waterford 50.59 Program is the same as that
applied by the industry guidance in NSAC-125 and is consistent with that previously
considered acceptable by the NRC staffin 1989. The margin of Safety is primarily focused on
those safety limits and protective boundaries which ensure the health and safety of the public.
Other margins that may be impacted involve design or operating margins that are within the
control of Waterford's design control processes.
.
6
1
,
3/20/98
Figure 1
Margin of Safety Determination Flowchart
Step 1
Safety Limit or
Protect Boundary
Nc
otentially Affecte
Yes
Step 2
g
identify Margins
(Explicit or Implicit)
that May Be Affected
in NRC SAR/SERs
o
"
Step 3
Step 4
Does Propose
No Margin of Safety
Change Affect the
Document
Exists
"
identified Margin of
Results
' Safety?
STOP PROCESS
Yes
u
Proposed Change
Document and
Represents a
- Request NRC Review
Margin of Safety
and Approval
per 10CFR50.59
7
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